Fuel elements for water-cooled nuclear reactors generally consist of sintered uranium or plutonium oxide pellets stacked inside a tubular cladding made from an alloy such as Zircaloy 2 or Zircaloy 4' which are zirconium-based alloys containing tin.
The cladding of the fuel element, among other things, enables direct contact to be prevented between the coolant of the reactor and the sintered fuel pellets.
The cladding material must therefore resist corrosion by the coolant which, in the case of pressurized-water nuclear reactors, consists of water containing various chemical additives, and must also resist a temperature of the order of 315.degree. C. and pressure of 155 bar. The cladding material must also have a low neutron absorption capacity.
Moreover, the sintered nuclear fuel pellets undergo various transformations when irradiated, in particular an expansion which may be responsible for a high contact pressure between certain fuel pellets and the inner surface of the cladding. The cladding is consequently subjected to mechanical stresses and is exposed to corrosive products created during the transformation of the fuel under irradiation, in the contact zones between the cladding and the pellets. These phenomena taken as a whole are generally referred to as pellet/cladding interaction (PCI).
In order to limit damage to the cladding caused by pellet/cladding interaction, it has been proposed to produce the cladding in a composite form such that it has an inner layer insulating the zirconium alloy tube from the fuel pellets.
Such a process is expensive and difficult to implement.
It has also been proposed to form various protective coatings inside the zirconium alloy cladding using conventional processes such as metal electroplating. These processes are difficult to implement in the case of tubes whose diameter is close to 9 mm and whose length is greater than 4 meters. Furthermore, the strength and efficiency of the coatings so formed more often than not prove to be insufficient when used in the environment of a nuclear reactor.
Coating-forming processes are also known, which processes are known under the name of chemical vapor deposition, or CVD. In the case where these processes are used for depositing a metal oxide layer onto a substrate, a volatile compound of the metal is brought into contact, in the presence of an oxidizing gas, with the substrate kept at a temperature permitting the in situ oxidation of the metal compound.
It has thus been proposed to deposit aluminum oxide, zirconium oxide, silica, or chromium or hafnium oxide using a halide, and generally a chloride, as the gaseous compound.
In order to implement these processes, it is generally necessary to keep the substrate at a high temperature of the order of 1000.degree. C. This requirement may have certain disadvantages in the case of cladding tubes for nuclear reactor fuel elements because of the sensitivity of the zirconium to halides and because of the temperature which alters the metallurgical state of the base metal.
It has also been proposed to deposit iron, indium or tin oxide, or alternatively to deposit zirconium oxide using an organometallic compound as the gaseous compound. In this case, the substrate may be brought to a temperature considerably less than 1000.degree. C.
In all cases, however, the depositing of oxide onto the substrate is a relatively slow process and the coating becomes extremely fragile as soon as the deposited layer exceeds a certain thickness, of the order of 1 to 2 microns.
It is therefore impossible to obtain, using the known processes of the prior art, deposits which are sufficiently thick and resistant to ensure effective protection against pellet/cladding interaction in the fuel elements of pressurized-water nuclear reactors.
Such deposits may advantageously consist of chromium oxide in the form of chromite Cr.sub.2 O.sub.3.
Chemical vapor deposition processes using an organometallic compound are known which may be carried out at a moderate temperature considerably less than 1000.degree. C. However, such processes have never been used to produce a chromium oxide insulating layer between the pellets and the cladding of a fuel element for a nuclear reactor.